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Journal Articles

Irradiation hardening in F82H irradiated at 573 K in the HFIR

Hirose, Takanori; Okubo, Nariaki; Tanigawa, Hiroyasu; Ando, Masami; Sokolov, M. A.*; Stoller, R. E.*; Odette, G. R.*

Journal of Nuclear Materials, 417(1-3), p.108 - 111, 2011/10

 Times Cited Count:13 Percentile:74.14(Materials Science, Multidisciplinary)

This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels irradiated at 573 K. The materials used in this research were F82H-IEA and its modified steels. Post irradiation mechanical tests revealed that irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated proof stress is less than 1 GPa. The deterioration of total elongation was also saturated by 9 dpa. Irradiation response of F82H-mod3, which is stable to temperature instability during material production and HIP treatment, was very similar to that of F82H-IEA, and negative impacts of extra tantalum was not observed. Therefore it can be an attractive option for the structural materials for blanket components manufactured by HIP.

Journal Articles

Tensile test technique for composites using small notched specimens

Nozawa, Takashi; Tanigawa, Hiroyasu

Journal of Nuclear Materials, 417(1-3), p.440 - 444, 2011/10

 Times Cited Count:5 Percentile:42.62(Materials Science, Multidisciplinary)

Development of small specimen test technique (SSTT) is one of key issues for irradiation study of nuclear-grade materials. This study aims to propose a new tensile test methodology using miniature notched specimens. For that purpose, crack extension behaviors of notched tensile specimens and their size effects were initially evaluated. Observing crack propagation behaviors of SiC/SiC composites, the apparent notch insensitivity was identified even if the moderate fiber/matrix interface was formed. Due to the notch insensitivity, it was found that key tensile properties can easily be predicted by test results of notched specimens. Noteworthy, this relation appeared independent of specimen size. Based on these findings, a conceptual design of miniature tensile test methodology using a notched specimen was proposed.

Journal Articles

Radiation hardening and IASCC susceptibility of extra high purity austenitic stainless steel

Ioka, Ikuo; Ishijima, Yasuhiro; Usami, Koji; Sakuraba, Naotoshi; Kato, Yoshiaki; Kiuchi, Kiyoshi

Journal of Nuclear Materials, 417(1-3), p.887 - 891, 2011/10

 Times Cited Count:7 Percentile:53.57(Materials Science, Multidisciplinary)

Fe-25Cr-35Ni EHP alloy was developed with conducting the countermeasure for IASCC. It is composed to adjust major elements, to remove harmful impurities and so on. The specimens were irradiated at 553 K for 25000h using JRR-3. The fluence was estimated to be 1.5$$times$$10$$^{25}$$n/m$$^2$$. Type 304SS was also irradiated as a comparison material. SSRT test was conducted in oxygenated water at 561 K in 7.7 MPa. The fracture mode of EHP alloy was ductile. IGSCC was not observed in the fracture surface. On the other hand, the fraction of IGSCC on the fracture surface of type 304 was about 70%. Microstructural evolution of EHP and type 304 after irradiation was examined by TEM. The defects induced by irradiation mostly consisted of black dots and frank loops in both specimens. No void was also observed in grain and grain boundary of both specimens. There was a little difference in microstructure after irradiation. It is believed that EHP alloy is superior to type 304 in irradiation.

Journal Articles

Basic technology for $$^{6}$$Li enrichment using an ionic-liquid impregnated organic membrane

Hoshino, Tsuyoshi; Terai, Takayuki*

Journal of Nuclear Materials, 417(1-3), p.696 - 699, 2011/10

 Times Cited Count:29 Percentile:91.36(Materials Science, Multidisciplinary)

The tritium as a fuel for fusion reactors is produced by the reaction of lithium-6 ($$^{6}$$Li) with a neutron in tritium breeding material. However, natural Li contains only about 7.6% $$^{6}$$Li, and the enrichment of $$^{6}$$Li up to 30 - 90% is required for tritium breeding material in the fusion reactor. The mercury amalgam method is superior as one of the lithium isotope enrichment methods, and might be currently utilized in practice. In Japan, on the other hand, the development of lithium isotope enrichment methods using the ion exchange membrane and molten salt has been conducted to avoid the environmental pollution. However, the isotope separation coefficient and efficiency is too low in the case of these methods. Therefore, these methods were difficult to apply to mass production for the large need of fusion blanket. Preliminary experiments were conducted. Organic membranes impregnated with TMPA-TFSI and PP13-TFSI as ionic liquids were prepared, and the relationship between the $$^{6}$$Li separation coefficient and the applied dialysis electric current was measured. The results showed that $$^{6}$$Li isotope separation coefficient by this method (about 1.2$$sim$$1.4) was larger than that by the mercury amalgam method (about 1.06).

Journal Articles

Development of advanced tritium breeding material with added lithium for ITER-TBM

Hoshino, Tsuyoshi; Kato, Kenichi*; Natori, Yuri*; Oikawa, Fumiaki; Nakano, Natsuko*; Nakamura, Mutsumi*; Sasaki, Kazuya*; Suzuki, Akihiro*; Terai, Takayuki*; Tatenuma, Katsuyoshi*

Journal of Nuclear Materials, 417(1-3), p.684 - 687, 2011/10

 Times Cited Count:39 Percentile:95.49(Materials Science, Multidisciplinary)

Lithium titanate (Li$$_{2}$$TiO$$_{3}$$) is one of the most promising candidates among tritium breeding materials because of its good tritium release. Addition of H$$_{2}$$ to inert sweep gas has been proposed for enhancing the tritium release from tritium breeding materials. However, the mass of Li$$_{2}$$TiO$$_{3}$$ was decreased with time in the hydrogen atmosphere. It is assumed that the mass decrease indicates the loss of the oxygen contained in the sample caused by the change from Ti $$^{4+}$$ to Ti $$^{3+}$$, and that the partial pressures of Li-containing species were increased in the hydrogen atmosphere. In order to decrease the mass-change at high temperature, advanced tritium breeding material with added Li should be developed to improve the physical and chemical stability in hydrogen atmosphere. In the case of the Li$$_{2+x}$$TiO$$_{3+y}$$ samples used by the present study, LiOHH$$_{2}$$O and H$$_{2}$$TiO$$_{3}$$ were proportionally mixed with the molar ratio Li/Ti of either 2.0 and 2.2. These samples are designated as L20 (Li/Ti = 2.0) and L22 (Li/Ti = 2.2), respectively. The results of XRD measurement showed that the phases in advanced tritium breeding material were as follows. L22 existed as non-stoichiometric compound Li$$_{2+x}$$TiO$$_{3+y}$$.

Journal Articles

Sintering properties of beryllides for advanced neutron multiplier

Nakamichi, Masaru; Yonehara, Kazuo

Journal of Nuclear Materials, 417(1-3), p.765 - 768, 2011/10

 Times Cited Count:40 Percentile:95.7(Materials Science, Multidisciplinary)

Journal Articles

Performance of various hydrophobic coatings to reduce HTO contamination

Iwai, Yasunori; Kobayashi, Kazuhiro; Yamanishi, Toshihiko

Journal of Nuclear Materials, 417(1-3), p.1187 - 1190, 2011/10

 Times Cited Count:2 Percentile:20.43(Materials Science, Multidisciplinary)

The concept of tritium containment and confinement is the root of fusion safety. Hence, HTO contamination on concretes and epoxy paint should be reduced as low as possible. Several kinds of hydrophobic coatings, a commercial silicic paint, a commercial acrylic paint, a commercial fluoric paint, methoxytrimethylsilane paint or metallic stick-sheets, were tested on concrete and epoxy samples. These samples were exposed to 740-1110Bq/cm$$^{3}$$ of HTO vapor at room temperature for a given period from 1 to 60 weeks. Static leaching tests were carried out for every HTO absorbed sample in distilled water, and the amount of leached HTO was evaluated. The hydrophobic barriers were effective to reduce HTO penetration into concrete. After exposure to HTO for 1 week, the HTO amount penetrated into concrete was reduced to 54.2% of non-paint sample for methoxytrimethylsilane paint, 56.0% for a commercial fluoric paint, 66.8% for a commercial silicic paint, respectively. Effectiveness of these hydrophobic barriers became less as the samples were exposed to HTO for a longer period.

Journal Articles

Recent progress in blanket materials development in the Broader Approach Activities

Nishitani, Takeo; Tanigawa, Hiroyasu; Nozawa, Takashi; Jitsukawa, Shiro; Nakamichi, Masaru; Hoshino, Tsuyoshi; Yamanishi, Toshihiko; Baluc, N.*; M$"o$slang, A.*; Lindou, R.*; et al.

Journal of Nuclear Materials, 417(1-3), p.1331 - 1335, 2011/10

 Times Cited Count:12 Percentile:71.79(Materials Science, Multidisciplinary)

As a part of the Broader Approach (BA) activities, the research and development on blanket related materials and tritium technology have been initiated toward DEMO by Japan and EU. Recently, those five R&D items have progressed substantially in Japan and EU. As a preparatory work aiming at the RAFM steel muss-production development, a 5-ton heat of RAFM steel (F82H) was procured with the Electro Slag Re-melting as a secondary melting. The result of the double notch tensile test method for the NITE-SiC$$_{f}$$/SiC specimen indicated notch insensitivity and very minor size effect on proportional limit tensile stress and fracture strength. For the fabrication technology development of beryllide neutron multiplayer pebbles, Be- Ti inter-metallic pebbles have been sintered directly from the mixed powder of Be and Ti in Japan.

Journal Articles

IFMIF's new design; Status after 2 years of EVEDA project

Garin, P.*; Sugimoto, Masayoshi

Journal of Nuclear Materials, 417(1-3), p.1262 - 1266, 2011/10

 Times Cited Count:14 Percentile:76.2(Materials Science, Multidisciplinary)

IFMIF is a major installation in the fusion programme to irradiate and characterize the fusion materials necessary for development of DEMO and the future fusion power plants. The Engineering Validation and Engineering Design Activities launched in mid 2007 in the framework of the Broader Approach agreement between EURATON and Japan comprise four sub-projects: (1) to complete engineering design of IFMIF, (2) to build and operate prototype of low energy part of accelerator up to 9MeV with 125mA beam current, (3) to validate full liquid lithium loop including purification traps and monitoring devices, and (4) to design and manufacture high flux test module of the test cell with testing in relevant conditions. Two years after the official start of the project the most important modifications to the reference design were high energy part of accelerator, lithium target assembly backplate, and high flux test module geometry. The impacts of these changes on the project are summarized.

Journal Articles

Characterization of the microstructure of dual-phase 9Cr-ODS steels using a laser-assisted 3D atom probe

Nogiwa, Kimihiro; Nishimura, Akihiko; Yokoyama, Atsushi; Otsuka, Satoshi; Kaito, Takeji; Inoue, Masaki; Okubo, Tadakatsu*; Hono, Kazuhiro*

Journal of Nuclear Materials, 417(1-3), p.201 - 204, 2011/10

 Times Cited Count:7 Percentile:53.57(Materials Science, Multidisciplinary)

Du se 9Cr-ODS (oxide dispersion-strengthened) steel consisting of residual-$$alpha$$ ferrite and $$alpha$$ prime martensite has excellent high-temperature strength. This study describes the microstructure of dual-phase 9Cr-ODS steels characterized by atom-probe tomography in order to compare oxide-particle dispersion states in each phase. This revealed that nano-size oxide particles were of the same chemical composition and that their mean size was about 3 nm in each phase. On the other hand, the number density in the residual-$$alpha$$ phase was about four times higher than that of the $$alpha$$ prime phase. These results indicate that the dense distribution of the oxide particles in the residual-$$alpha$$ phase contribute to the excellent high-temperature strength of 9Cr-ODS steel.

Journal Articles

Density functional calculations for small iron clusters with substitutional phosphorus

Nakazawa, Tetsuya; Igarashi, Takahiro; Tsuru, Tomohito; Kaji, Yoshiyuki; Jitsukawa, Shiro

Journal of Nuclear Materials, 417(1-3), p.1090 - 1093, 2011/10

 Times Cited Count:3 Percentile:28.8(Materials Science, Multidisciplinary)

It is well known that impurities in iron which segregate to grain boundaries can dramatically change physical and chemical properties. Phosphorous, which is one of impurities, segregates at grain boundaries under thermal or irradiation environments, and brings about the intergranual embrittlement. In this study, influences of phosphorus substitutions for binding energies and electronic structures of octahedral iron cluster are investigated computationally using density functional calculations in order to understand the nature of bonding between phosphorus and iron at grain boundaries. The values of binding energies of cluster are increasing with the phosphorous substitutions. The increases are due to Fe-P bond strengthened by the charge transfer from Fe atom to P atom. On the other hand the calculated bond orders give that Fe-Fe bonds are weakened. Thus, the embrittlement induced with the segregation of P due to irradiation is considered to be associated with weakened Fe-Fe bonds.

Journal Articles

Modeling of the grain boundary segregation of helium in $$alpha$$-Fe

Suzudo, Tomoaki; Kaburaki, Hideo; Yamaguchi, Masatake

Journal of Nuclear Materials, 417(1-3), p.1102 - 1105, 2011/10

 Times Cited Count:9 Percentile:62.21(Materials Science, Multidisciplinary)

This paper discusses a computational approach to the helium segregation at grain boundaries with reference to the helium embrittlement under the fast neutron irradiation; we particularly applied a kinetic Monte-Carlo method to the modeling of the segregation phenomena in alpha-iron. In the calculation, the migration and segregation energies of helium atoms given by the first principle calculations are used. Obtained results are discussed in comparison with McLean's equilibrium segregation theory.

Journal Articles

Effect of welding and coating on deuterium permeation through F82H

Nakamura, Hirofumi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Yamanishi, Toshihiko

Journal of Nuclear Materials, 417(1-3), p.1150 - 1153, 2011/10

 Times Cited Count:3 Percentile:28.8(Materials Science, Multidisciplinary)

Deuterium permeation characteristics through various surface states ofF82H such as F82H without surface treatment (bare F82H), welded F82H,and gold plated F82H (Au-F82H) have been investigated in order to understand the effect of surface state on permeation for the realistic tritium permeation evaluation in the fusion reactors especially breeding blanket system, which has many welding points and permeation reduction coatings. Based on obtained permeation behavior, the steady state permeation and the diffusivity derived form the transient permeation behavior have been discussed. As the results, deuterium permeation through bare F82H is smaller than that through clean surface F82H. Asto the effect of welding on permeation, the results indicate that significant difference of welding is not observed between bare F82H and welded F82H. Finally, gold plating on F82H showed good permeation reduction performance.

Journal Articles

Study of the behavior of tritiated water vapor on concrete materials

Kobayashi, Kazuhiro; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko

Journal of Nuclear Materials, 417(1-3), p.1183 - 1186, 2011/10

 Times Cited Count:2 Percentile:20.43(Materials Science, Multidisciplinary)

In a fusion reactor of high safety and acceptability, safe confinement of tritium is one of key issues for the fusion reactor. Tritium should be well-controlled and not excessively released to environment and to prevent workers from excess exposure. Especially, the hot cell and tritium facility of ITER will be used various construction materials such as the concrete, the organic materials. As the results, the concrete materials were almost saturated with HTO vapor within about 1month except for cement paste and it was larger in the order of cement paste $$>$$ mortar $$>$$ concrete. Even if one month passes from the exposure beginning, the amount of sorbed tritium to cement paste did not reach saturation. The chemical form of desorbed tritium from the sample was almost HTO. In addition, the tritium behavior that adsorbs the surface of concrete materials will be discussed by using FT-IR.

Journal Articles

Development of a low activation concrete shielding wall by multi-layered structure for a fusion reactor

Sato, Satoshi; Maegawa, Toshio*; Yoshimatsu, Kenji*; Sato, Koichi*; Nonaka, Akira*; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara

Journal of Nuclear Materials, 417(1-3), p.1131 - 1134, 2011/10

 Times Cited Count:7 Percentile:53.57(Materials Science, Multidisciplinary)

The multi-layered concrete structure has been developed to reduce induced activity in the concrete applied for neutron generation facilities such as a fusion reactor. The multi-layered concrete structure is composed of the low activation concrete as the first layer, the boron-doped low activation concrete as the second layer and the ordinary concrete as the third layer from the side of the neutron source. By applying the multi-layered concrete structure, the volume of the boron can drastically decrease compared with the monolithic boron-doped concrete. A 14 MeV neutron irradiation experiment with the multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activities in the mockups were measured. This experiment demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activities.

Journal Articles

Stability of non-stoichiometric clusters in the MA957 ODS ferrtic alloy

Sakasegawa, Hideo; Legendre, F.*; Boulanger, L.*; Brocq, M.*; Chaffron, L.*; Cozzika, T.*; Malaplate, J.*; Henry, J.*; de Carlan, Y.*

Journal of Nuclear Materials, 417(1-3), p.229 - 232, 2011/10

 Times Cited Count:57 Percentile:98.02(Materials Science, Multidisciplinary)

In our past work, the commercial ferrtic Oxide Dispersion Strengthened (ODS) alloy MA957 had at least two types of nanometer-sized oxide particles: non-stoichiometric Y-, Ti-, O-enriched clusters and Y$$_{2}$$Ti$$_{2}$$O$$_{7}$$ particles. The size of the non-stoichiometric clusters was much smaller than that of Y$$_{2}$$Ti$$_{2}$$O$$_{7}$$ particles and it was confirmed that the non-stoichiometric clusters possibly dominate the oxide dispersion strengthening. Therefore, this study dealt with the stability and evolution mechanisms of non-stoichiometric nanoclusters after the annealing (1473K $$times$$ 1h). This annealing condition was determined considering the actual condition of consolidation processes. After the annealing, most non-stoichiometric Y-, Ti-, O-enriched clusters were stable, but some clusters became Y$$_{2}$$Ti$$_{2}$$O$$_{7}$$ particles with increasing size. The diffusion of yttrium had an important role for the evolution of these oxides.

Journal Articles

Present status of refurbishment and irradiation technologies in JMTR

Inaba, Yoshitomo; Ishihara, Masahiro; Niimi, Motoji; Kawamura, Hiroshi

Journal of Nuclear Materials, 417(1-3), p.1348 - 1351, 2011/10

 Times Cited Count:3 Percentile:20.43(Materials Science, Multidisciplinary)

The Japan Materials Testing Reactor (JMTR) is a testing reactor with first criticality in March 1968. JMTR has been utilized for various neutron irradiation tests on nuclear fuels and materials, as well as for radioisotope production. The operation of JMTR was stopped in August 2006 for the refurbishment and the improvement. The renewed JMTR will be operated from FY 2011. Aiming at the restart of the new JMTR, the renewal of the aging reactor components, the preparation of the new irradiation facilities, and the development of the irradiation technologies have been carried out in JMTR. The irradiation facilities and technologies can also contribute to the development of fusion reactor materials. In this paper, the present status of the refurbishment and the irradiation technologies focused on the instrumentation in JMTR are described.

Journal Articles

Theoretical study on segregation of Cu, Mo and W impurities and stability of impurity-vacancy pairs in bcc Fe

Tsuru, Tomohito; Abe, Yosuke; Suzuki, Chikashi; Nakazawa, Tetsuya; Kaji, Yoshiyuki; Tsukada, Takashi

Journal of Nuclear Materials, 417(1-3), p.1054 - 1057, 2011/10

 Times Cited Count:1 Percentile:11.72(Materials Science, Multidisciplinary)

Reduced activation ferritic steel is one of the leading structural material candidates for a nuclear fusion reactor. Since the solute impurities have an effect on the embrittlement through segregation under irradiation, the stability of impurity elements should be elaborated. In the present study the segregation characteristics of tungsten and some general solute impurities in bcc iron were investigated nonempirically by first principle calculations, where the equilibrium segregation was considered via a regular solution model and the change in enthalpy for segregation were directly evaluated for comparison. Subsequently the energetic stabilities of impurity-impurity and impurity-vacancy pair were evaluated. The segregation enthalpy is influenced by the electronic interaction between the d-electron of Fe and the outer electron of the impurity element, and molybdenum and tungsten tend to prevent from the impurity segregation.

Journal Articles

Flexible heat resistant neutron shielding resin

Sukegawa, Atsuhiko; Anayama, Yoshimasa*; Okuno, Koichi*; Sakurai, Shinji; Kaminaga, Atsushi

Journal of Nuclear Materials, 417(1-3), p.850 - 853, 2011/10

 Times Cited Count:20 Percentile:85.03(Materials Science, Multidisciplinary)

A flexible heat resistant neutron shielding material has been developed, which consists of polymer resin with 1 weight % boron. The neutron shielding performance of the developed resin, examined by the $$^{252}$$Cf neutron source is almost the same as that of the polyethylene. The outgas of H$$_{2}$$, H$$_{2}$$O, CO and CO$$_{2}$$ from the resin have been measured at 250 $$^{circ}$$C environment. The resin will be applied around the port of the vacuum vessel as an additional shielding material and prevented the effects on the neutron streaming of the superconducting tokamak device such as JT-60SA.

Journal Articles

SCC susceptibility of cold-worked stainless steel with minor element additions

Nakano, Junichi; Nemoto, Yoshiyuki; Tsukada, Takashi; Uchimoto, Tetsuya*

Journal of Nuclear Materials, 417(1-3), p.883 - 886, 2011/10

 Times Cited Count:1 Percentile:11.72(Materials Science, Multidisciplinary)

To examine the effects of minor elements on stress corrosion cracking (SCC) susceptibility of low carbon stainless steels with work hardened layer, a high purity type 304 stainless steel was fabricated and minor elements, Si, S, P, C or Ti, were added. Work hardened layer was introduced by shaving on the surface of stainless steels. The specimens were exposed to a boiling 42% MgCl$$_{2}$$ solution for 20 hours and the number and the length of initiated cracks were examined. SCC susceptibility of the specimen with P was the highest and that of the specimen with C was the lowest in all specimens. By magnetic force microscope examination, magnetic phase expected to be martensitic phase was detected near surface. Since corrosion resistance of martensite is lower than that of austenite, the minor elements additions would affect SCC susceptibility through the amount of the transformed martensite, i.e. austenite stability.

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